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In the framework of the Generation IV research and development project, in which the French Commission of Alternative and Atomic Energies (CEA) is involved, a main objective for the design of Sodium-cooled Fast Reactor (SFR) is to meet the safety goals for severe accidents. Among the severe ones, the Unprotected Transient OverPower (UTOP) accidents can lead very quickly to a global melting of the core. UTOP accidents can be considered either as slow during a Control Rod Withdrawal (CRW) or as fast. The paper focuses on fast UTOP accidents, which occur in a few milliseconds, and three different scenarios are considered: rupture of the core support plate, uncontrolled passage of a gas bubble inside the core and core mechanical distortion such as a core flowering/compaction during an earthquake. Several levels and rates of reactivity insertions are also considered and the thermal-mechanical behavior of an ASTRID fuel pin from the ASTRID CFV core is simulated with the GERMINAL code. Two types of fuel pins are simulated, inner and outer core pins, and three different burn-up are considered.Moreover, the feedback from the CABRI programs on these type of transients is used in order to evaluate the failure mechanism in terms of kinetics of energy injection and fuel melting. The CABRI experiments complete the analysis made with GERMINAL calculations and have shown that three dominant mechanisms can be considered as responsible for pin failure or onset of pin degradation during ULOF/UTOP accident: molten cavity pressure loading, fuel-cladding mechanical interaction (FCMI) and fuel break-up.The study is one of the first step in fast UTOP accidents modelling with GERMINAL and it has shown that the code can already succeed in modelling these type of scenarios up to the sodium boiling point. The modeling of the radial propagation of the melting front, validated by comparison with CABRI tests, is already very efficient.

AbstractThe main objective of the current study is to establish a justifiable data-driven students-faculty ratio (SFR) benchmark for higher education institutions (HEIs) in the Sultanate of Oman. This is established based on analyzing selected secondary data from three sources using purposive sampling. This study reflects quantitative research adopting descriptive analysis (frequency and percentage). The three sources of data used to arrive at this national benchmark are: (1) World Bank Open Data, (2) Quacquarelli Symonds (QS) University Rankings, and (3) THE (Times Higher Education) University Rankings. Based on 13 data points, the simple (unweighted) average SFR for Oman is 19.04. The current study recommends an upper limit of 30 students per academic full time faculty member in Omani HEIs. When the most-recent records in the World Bank Open Data portal for the 22 Arab League countries were averaged, a regional benchmark of 24.82 was obtained (which decreases to 17.00 if a subset of the six Gulf Cooperation Council, or GCC, countries is considered). The seven HEIs (all are universities) that received a rank in the latest edition (2024 edition) of QS Arab Region University Rankings (QS-AUR) were among the data points considered in the analysis. These QS ranking results were published online on 18/October/2023, with a total of 223 ranked Arab HEIs. The five HEIs (four universities and one standalone university college) that either received a rank (only Sultan Qaboos University, or SQU) or were recognized as a reporter without being qualified enough to officially receive a rank in the latest edition (2024 edition) of THE World University Rankings (THE-WUR) were also considered in the analysis. These THE ranking results were published online on 27/September/2023, with a total of 1,904 ranked HEIs worldwide. As a secondary objective, the study assesses the level of mismatch in the SFR value for four Omani universities that appeared in the 2024 edition of both QS-AUR and THE-WUR. The percentage of mismatch in SFR between the QS and THE values (with respect to the average of the two values) for each of these four universities ranged from 0.68% to 25.30%. As a tertiary objective, the study gives an overview about three popular university ranking services, which are: QS, THE, and ARWU (Academic Ranking of World Universities, also known as Shanghai Ranking). Sultan Qaboos University (SQU) is the only Omani HEI that received a global rank through either the QS or THE ranking service in their respective latest edition (edition 2024, released in 2023), and it is the only Omani HEI that was ever ranked by ARWU since its first edition in June 2003 (but it was not ranked in the latest ‘2023’ edition). The enrollment of Arab HEIs in the regional Arab version of THE rankings has grown faster than the regional QS counterpart when the latest 3 released editions are considered for both types of regional rankings, where the number of Arab HEIs ranked in the 2023 edition (released on 15/November/2023) of the Arab version of THE rankings (having 207 ranked Arab HEIs) is notably higher than its value in the earlier 2021 edition released on 27/July/2021 (having 125 ranked Arab HEIs).

To evaluate the predictive ability of the RIRS scoring system and the RUSS in predicting stone-free rate (SFR) after retrograde intrarenal surgery (RIRS). This prospective study was conducted on patients who underwent RIRS for kidney stones. Two scoring systems were used to determine the degree of procedure difficulty: the RIRS scoring system and the RUSS. We assessed the predictive ability of the two scoring systems utilizing receiver operating characteristic (ROC) analysis and calculated the sensitivity and specificity of each system. Additionally, we analyzed the association between the scoring systems and the stone-free outcome using a multivariate logistic regression model. One hundred seventy-one patients were incorporated into this study with a mean age of 43 years, and 65.5% were male. The results showed a significant AUC of 0.868 for the RIRS score (P < 0.001, 95% CI = 0.813–0.924). The sensitivity and specificity were 72% and 93.7%, respectively. In contrast, the RUSS score revealed a non-significant unsatisfactory AUC of 0.480 (P = 0.660), with a 95% confidence interval ranging from 0.384–0.576. The RIRS scoring system showed a better predictive ability for SFR after RIRS than the RUSS. Additionally, RIRS was a significant predictor of SFR, controlling for age, gender, body mass index, and previous renal surgery. [ABSTRACT FROM AUTHOR]

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Sodium cooled Fast Reactors (SFR) are built with several engineered safety features and hence a severe accident such as a core melt accident is hypothetical with a probability of

The main goal of the Vera C. Rubin observatory is to perform the 10 year Legacy Survey of Space and Time (LSST). This future state-of-art observatory will open the new window to study billions of galaxies from Local Universe as well as the high redshift objects. In this work we employ simulated LSST observations and uncertainties, based on the 50 385 real galaxies within the redshift range 0 < z < 2.5 from the ELAIS-N1 and COSMOS fields of the Herschel Extragalactic Legacy Project (HELP) survey, to constrain the physical properties of normal star-forming galaxies, such as their star formation rate (SFR), stellar mass (Mstar), and dust luminosity (Ldust). We fit their spectral energy distributions (SEDs) using the Code Investigating GALaxy Emission (CIGALE). The stellar masses estimated based on the LSST measurements agree with the full UV to far-IR SED, while we obtain a clear overestimate of the dust-related properties (SFR, Ldust) estimated with LSST. We investigate the cause of this result and find that it is necessary to employ auxiliary rest-frame mid-IR observations, simulated UV observations, or the far-UV attenuation (AFUV)-Mstar relation to correct for the overestimate. [ABSTRACT FROM AUTHOR]

• The accuracy of the classification of clavicle fractures in the Swedish Fracture Register was fair. • The Robinson clavicle fracture classification had almost perfect inter- and intraobserver agreement among the expert raters. • In the Swedish Fracture Register clavicle fractures with only partial displacement were often incorrectly classified as fully displaced. • In order to improve accuracy, the classification instructions for clavicle fractures in the Swedish Fracture Register need to be updated. In the Swedish Fracture Register (SFR) clavicle fractures are classified according to the Robinson classification. This study aimed to evaluate the accuracy of the classification of clavicle fractures in the SFR. A secondary aim was to assess inter- and intraobserver agreement. Clavicle fractures (n = 132) were randomly selected from the SFR and radiographs were requested for each patient from their treating departments. Not all radiographs could be acquired, and after exclusion, 115 fractures were independently classified by three expert raters blinded to patient information. The 115 fractures were classified on two occasions, 3 months apart. The raters' consensus classification was used as a gold standard that was compared to the classification registered in the SFR. The accuracy, defined as the degree of agreement between the gold standard and SFR classifications, was reported, as was the inter- and intraobserver agreement for the expert raters. Agreement between the classification in the SFR and the gold standard classification was fair (kappa = 0.35). Fractures with only partial displacement were often incorrectly classified as fully displaced in the SFR (n = 31 of 78 displaced fractures in the SFR). The inter- and intraobserver agreement among the expert raters was almost perfect (interobserver kappa = 0.81–0.87, intraobserver kappa = 0.84–0.94). The accuracy of the classification of clavicle fractures in the SFR was only fair, whereas the inter- and intraobserver agreement among the expert raters was almost perfect. Accuracy in the SFR may be improved if the classification instructions in the SFR are updated by incorporating the original classification displacement criteria, both in text and in illustrated form. [ABSTRACT FROM AUTHOR]

BackgroundMany parents develop stress-related symptoms and depression when their preterm infant is hospitalised in the neonatal intensive care unit (NICU) after birth. We reviewed the evidence of parent well-being with preterm infants hospitalised in single family rooms (SFRs) or in open bay neonatal units (OBUs).MethodsFor this systematic review and meta-analysis, we searched MEDLINE, EMBASE, PsycINFO, Cochrane Central Register of Controlled Trials (CENTRAL), Web of Science, Clinicaltrials.gov, and International Clinical Trials Registry Platform (ICTRP) databases from inception through 22 November 2019 using controlled terms and text words related to prematurity and NICU-design. We included randomised and non-randomised studies comparing outcomes in parents with preterm infants admitted to SFRs or OBUs. Methodological quality was assessed using Cochrane Collaboration's Risk of Bias Tool for randomised controlled trials and the Risk of Bias Tool for Non-Randomised Studies of Interventions (ROBINS-I). Outcomes included: parental stress, satisfaction, participation (presence/involvement/skin-to-skin care), self-efficacy, parent-infant-bonding, depression, anxiety, post-traumatic stress, empowerment, and degree of family-centred care. Summary estimates were calculated using random effects models with standardised mean differences (SMDs). PROSPERO registration: CRD42016050643.FindingsWe identified 614 unique publications. Eleven study populations (1, 850 preterm infants, 1, 549 mothers and 379 fathers) were included. All but one study were at serious to critical risk of bias. SFRs were associated with higher levels of parental presence, involvement, and skin-to-skin care. Upon discharge, SFRs were associated with lower stress levels (n = 828 parents, SMD-0·30,95%CI -0·50;-0·09, p

The RVACS is a versatile and robust safety system driven by two natural circulations: in-vessel coolant and ex-vessel air. To observe interaction between the two natural circulations, SINCRO-IT facility was designed with two different similarity laws simultaneously. Bo’ based similarity law was employed for the in-vessel, while Ishii’s similarity law for the ex-vessel excluding the radiation. Compared to the prototype, the sodium and air system, SINCRO-IT was designed with Wood’s metal and air, having 1:4 of the length reduction, and 1.68:1 of the time scale ratio. For the steady state, RV temperature limit was violated at 0.8% of the decay heat, while the sodium boiling was predicted at 1.3%. It showed good accordance with the system code, TRACE. For an arbitrary re-criticality scenario with RVACS solitary operation, sodium boiling was predicted at 25,100 s after power increase from 1.0 to 2.0%, while the system code showed 30,300. Maximum temperature discrepancy between the experiments and system code was 4.2%. The design and methodology were validated by the system code TRACE in terms of the convection, and simultaneously, the system code was validated against the simulating experiments SINCRO-IT. The validated RVACS model could be imported to further accident analysis.

The decoupling relationship and low inertia characteristics of wind turbine generator system (WTGS) make the system frequency response more complex. In order to evaluate the frequency stability of high-penetration wind power grid, it is significant to establish an exactly frequency response analysis model. Due to the features of wide spatial distribution and large number of WTGS in the power grid, the wind speed and actual operating state faced by WTGS in different regions have obvious discrepancy, which makes the traditional frequency response models that merely considers the single operating state of WTGS have evident limitations. Therefore, in order to consider the discrepancy of frequency regulation characteristics caused by the spatial dispersion of WTGS, an improved frequency response modeling method which can take into account the frequency regulation characteristics of multiple wind speeds is proposed. This method uses small signal analysis theory and fuzzy control method to construct an improved equivalent aggregation model of WTGS. Combining this model with the typical system frequency response (SFR) model, a novel frequency response model considering different wind conditions and different frequency regulation characteristics of WTGSs is established. Finally, an improved IEEE-118 bus system is used for case analysis to verify the correctness and superiority of the proposed method for high-penetration wind power grid.

Abstract Striving to peak carbon emissions and achieve carbon neutrality (known as the "Dual-Carbon" goal) is an inevitable requirement for elevating the environmental resource constraints and realizing harmonious coexistence between the mankind and the earth. In the energy system, nuclear energy offers various advantages, such as high energy density, low carbon emission, strong environmental adaptability and large potential for energy co-generation and co-supply. It is one of the supporting energy sources for the transformation and upgradation of the energy system to a clean, efficient and low-carbon way. In this paper, the opportunities and challenges for innovation-driven nuclear energy development in the fields of electricity generation, hydrogen production, heat supply and seawater desalination under the goal of "Dual-Carbon" are discussed and analyzed. Besides, the relevant research on improving the safety and economy of the pressurized water reactor (PWR) and sodium-cooled fast reactor (SFR) conducted by the Nuclear THermal–hydraulic research Lab (NuTHeL) of Xi’an Jiaotong University is briefly introduced.

Several countries show interest in the Generation-IV power reactor innovative small module (PRISM), including: Canada, Japan, Korea, Saudi Arabia and the United Kingdom. Generation IV International Forum (GIF) has recommended the utilizing of probabilistic safety assessment (PSA) in evaluating the safety of Generation-IV reactors. This paper reviews the PSA performed for PRISM using SAPHIRE 7.27 code. This work shows that the core damage frequency (CDF) of PRISM for a single module is estimated by 8.5E-8/year which is lower than the Generation-IV target that is 1E-6 core damage per year. The social risk of PRISM (likelihood of latent cancer fatality) with evacuation is estimated by 9.0E-12/year which is much lower than the basic safety objective (BSO) that is 1E-7/year. The social risk without evacuation is estimated by 1.2E- 11/year which is also much lower than the BSO. For the individual risk (likelihood of prompt fatality), it is concluded that it can be considered negligible with evacuation (1.0E-13/year). Assuming no evacuation, the individual risk is 2.7E-10/year which is again much lower than the BSO. In comparison with other PSAs performed for similar sodium fast reactors (SFRs), it shows that PRISM concept has the lowest CDF.

When a Core Disruptive Accident (CDA) occurs in a Sodium-cooled Fast Reactor (SFR), as the accident progresses, its core may form a pool of molten fuel. When the molten fuel pool expands, a certain amount of coolant may be entrained in it. Because of the Fuel-Coolant Interaction (FCI), the molten fuel pool will have centralized sloshing behavior, making the fuel distribution more dense, leading to the risk of re-criticality, therefore, the study of centralized sloshing behavior is of great significance for evaluating the impact of the CDA. Under various experimental conditions, the sloshing experiment of gas injection can effectively simulate the mechanism and characteristics of centralized sloshing behavior. In this study, numerical simulation of gas-injection induced sloshing behavior is performed based on the Moving Particle Semi-implicit (MPS) method. Simulation results are compared with the experimental results. The results show that the deformation of bubble shape and liquid level in the sloshing characteristics are in good agreement with the experimental results. Therefore, the numerical simulation method established in this study can be used to study the pool sloshing characteristics induced by gas injection.

Pronghorn-SC is a subchannel code within the Multiphysics Object-Oriented Simulation Environment (MOOSE). Initially designed to simulate flows in water-cooled, square lattice, subchannel assemblies, Pronghorn-SC has been expanded to simulate liquid-metal-cooled flows in triangular lattices, hexagonal subchannel assemblies. For this purpose, the algorithm of Pronghorn-SC was adapted to solve the subchannel equations as they are applicable to a hexagonal wire-wrapped sodium-cooled fast reactor. Cheng–Todreas models for pressure drop and cross-flow models were adopted and a coolant heat conduction term was added. To solve these equations, an improved implicit algorithm was developed robust enough to deal with the numerical issues, associated with low flow and recirculation phenomena. To confirm the prediction capability of Pronghorn-SC, calculations and comparisons with available experimental data of 19- and 37-pin assemblies were performed, as well as other subchannel codes. Finally, a flow blockage modeling feature was added. This capability was validated for both water-cooled square sub-assemblies and sodium-cooled hexagonal sub-assemblies, using experimental data of partially and fully blocked cases. [ABSTRACT FROM AUTHOR]

IntroductionTo predict the factors of residual stones after percutaneous nephrolithotomy (PCNL) by analyzing the characteristics of the renal anatomical structure in intravenous urography, so as to make a reasonable operation plan, reduce the risk of residual stones in PCNL, and improve the stone-free rate (SFR).MethodsA retrospective study was performed between January 2019 and September 2020 for patients treated with PCNL. According to the results of a kidney ureter bladder review after PCNL, 245 patients were divided into a residual stone group (71 patients, stone size >4 mm) and a stone-free group (174 patients, stone size ≤4 mm). An independent sample t-test was used to analyze the age, the length and width of channel calices, the angle between the channel calices and the involved calices, and the length and width of the involved calices. The gender, the channel types, the number of channels, the degree of hydronephrosis, and the number of involved calices were analyzed by using the chi-square test. A score of p

Introduction and Objectives. Flexible ureterorenoscopy (f-URS) gained a lot of field in treating reno-ureteral lithiasis, with a clear indication for kidney stones smaller than 20 mm with an acceptable stone free rate (SFR) compared to its complications, but also with a specific learning curve. The goal of this study is to investigate the learning curve in f-URS and its impact on success and complication rates. Materials and Methods. 60 patients from a tertiary care urology clinic who underwent f-URS were enrolled in this prospective study and divided into two study groups of 30 patients each based on the period of time the surgery was performed. Patients' and lithiasis characteristics, f-URS-related parameters, hospital stay time, SFR and complications were gathered into a database for comparison and statistical assessment of the differences. Results. The two study groups were comparable, regarding the age, sex, BMI, stone location, stone density and stone sizes. Time passed till the introduction of the flexible ureteroscope was 18.5 minutes in group 1 versus 12.4 minutes in group 2 (p = 0.001). The fluoroscopy time was shorter in group 2 40.8 versus 83 seconds (p = 0.001). The operative time decreased significantly from 49.2 minutes to 32 minutes in the second group (p = 0.001). Complication rate and hospital stay showed no major difference between the two patient groups, but the SFR was significantly improved in the latest study group-86.66% versus 66%. Conclusions. Flexible ureteroscopy is a minimally invasive surgical intervention that requires experience and surgical skills gained after overpassing a specific learning curve. Our study shows that a good level of proficiency might be obtained after the completion of at least 30 cases in terms of operative time, fluoroscopy time, complication, and stone free rates. Further standardized prospective trials ought to be conducted to better characterize assess the precise learning curve. [ABSTRACT FROM AUTHOR]

[Display omitted] • Fragmentation of simulated corium in sodium was studied using X-ray imaging system. • Fine fragments were generated, which found to solidify within short travel in sodium. • MFCI in sodium indicate a reduced potential for an energetic MFCI than that of water. • Absence of large size debris indicate favorable post-accident heat transfer in sodium. In case of a hypothetical core meltdown accident in Sodium cooled Fast Reactor (SFR), corium interacts with cold pool sodium and results in Molten Fuel Coolant Interaction (MFCI). During MFCI, the corium fragments and resultant debris settle on the core catcher as a bed and continue to generate decay heat. The size and state of fragmented debris play a crucial role in post-accident heat transfer. An experimental study has been undertaken to investigate the fragmentation of simulated corium in sodium using real-time X-ray imaging. Experiments were conducted by releasing ∼400 g of molten mixture of alumina and iron at ∼2400 °C into ∼4 kg of sodium at various temperatures. The fragmentation phenomenon was acquired using a digital flat panel detector and the captured images were subjected to image analysis to extract information on melt fragmentation. Post-experiment, fragmented debris was retrieved and subjected to sieve analysis to obtain characteristics towards assessing the dominant fragmentation mechanism. Experiments were also conducted with water for identifying key differences in the fragmentation and debris characteristics. Non-energetic fragmentation, generation of fine debris, and relatively minor sodium vaporization observed in the small-scale experiments present a reduced potential for energetic interaction in sodium than that of water. [ABSTRACT FROM AUTHOR]

Chromium (Cr) has been recognized as a promising diffusion barrier candidate to mitigate the Fuel Cladding Chemical Interaction (FCCI) failure in metallic fuels for sodium-cooled fast reactor. This paper, for the first time, conducted an in-depth post-irradiation examination of the microstructure and composition evolution, and micromechanics of the Cr diffusion barrier in U-10Zr fuel/HT9 cladding irradiated in the Advanced Test Reactor (ATR) to 8.7% burnup at an averaged Peak Inner Cladding Temperature of 540–550 °C. Transmission Electron Microscope (TEM) characterization confirmed the preferential intergranular diffusion of Zr and U in the Cr diffusion barrier, suggesting that grain boundaries served as fast path for the diffusion of Zr and U into the Cr diffusion barrier. The interaction zone is dominated by nano crystalline α- ZrCr 2 Laves phase. Despite the interaction, there is no microcracks being observed in the preserved Cr diffusion barrier and HT9 cladding, serving as a good barrier to mitigate FCCI under the studied in-reactor irradiation conditions. High density cavities in uniform distribution are observed inside Cr grains, nano particles contains Cr, Mn, and O are confirmed by Electron Energy Loss Spectroscopy (EELS) analysis. However, it is unclear whether the cavities will become an issue to the barrier integrity at higher burnup. In-situ Scanning Electron Microscopy (SEM) micro-tensile testing uncovered mechanical softening in the HT9 cladding nearing the Cr diffusion barrier, possibly due to the coarsening of lath structure and carbides precipitates. [ABSTRACT FROM AUTHOR]

• A new specimen has been designed that uses only a small amount of sodium (about 20 g) for this work. • The creep behavior of 316LN in oxygen-saturated liquid sodium at high temperatures (650 ℃, 700 ℃ and 750 ℃) is tested and compared with the results of creep tests in air at the same conditions. • The corrosion mechanism of 316LN under external loading in oxygen-saturated sodium is illustrated. The high-temperature performance and sodium corrosion of in-core components are the main problems of liquid sodium-cooled reactors, and low-carbon nitrogen-controlled 316 stainless steel (316LN) is the main research and development direction of the main materials for the core of sodium-cooled fast reactors. In this work, the creep behavior of 316LN in oxygen-saturated liquid sodium at high temperatures (650 ℃, 700 ℃ and 750 ℃) was tested and compared with the results of creep tests in air at the same conditions. It is found that the creep life in the sodium environment is reduced by 63.44%, 67.93% and 40.07%, respectively. The specimens in oxygen-saturated sodium exhibit reduction in ductility, selective dissolution as well as oxidative corrosion. The corrosion mechanism of 316LN under external loading in oxygen-saturated sodium is illustrated. [ABSTRACT FROM AUTHOR]

The downward discharge of the molten fuel to the lower structure of the fuel assembly could increase of the pressure drop and degrade of coolability of the assembly. To analyze the phenomena, experiments for the generation of the debris bed were conducted as LOF-DT series. Based on the debris bed in the LOF-DT, pressure drop experiment was conducted with intact and blocked component. Parametric study on the pressure drop was conducted by CFD. The LOF-DT experiments were conducted for the position and porosity of the debris bed. 85% of the debris were sedimented in the lower reflector, and 15% were in the nose piece, approximately. Porosity of the debris bed were about 0.7 and 0.85 in the lower reflector and nose piece, respectively. Pressure drop increased significantly with debris bed, especially in the lower reflector. More than 120 time of the pressure drop increased in the lower reflector, while only 10% increased in the nose piece. According to the parametric study, mass of the debris was the most important for pressure drop. The lower discharge phenomena could have a significant effect to the total pressure drop of the fuel assembly, approximately 10.8 times for the base case.

Sodium-cooled fast reactor (SFR) is the preferred technology of the generation-IV fast neutron reactor, and its core body mainly uses nuclear-grade 316 stainless steel. In order to prolong the design life of SFRs to 60 years and more, it is necessary to summarize and analyze the anti-corrosion effect of nuclear grade 316 stainless steel in high temperature sodium environment. The research on sodium corrosion of nuclear grade 316 stainless steel is mainly composed of several important factors, including the microstructure of stainless steel (ferrite layer, degradation layer, etc.), the trace chemical elements of stainless steel (Cr, Ni and Mo, etc) and liquid impurity elements in sodium (O, C and N, etc), carburization and mechanical properties of stainless steel, etc. Through summarizing and constructing the sodium corrosion rate equations of nuclear grade 316 stainless steel, the stainless steel loss of thickness can be predicted. By analyzing the effects of temperature, oxygen content in sodium and velocity of sodium on corrosion rate, the basis for establishing integrity evaluation standard of SFR core components with sodium corrosion is provided.

Global carbon emissions have been rising sharply since the start of the 20th century, and countries have adopted various policies in recent years to reduce greenhouse gas (GHG) emissions in different sectors. Nuclear energy is one energy source that is least polluting with minimum GHG emissions. India’s nuclear power programme started with Heavy water reactors in the first stage followed by Fast Reactors in the second stage. Third stage of Thorium utilisation is yet to start. The deployment of Pu/depleted U from Heavy water reactors in fast reactors would help in the effective utilisation of the indigenous uranium resources to a large extent besides reducing the waste. The thorium technology to obtain uranium 233 is equally important as India possesses large amounts of thorium deposits. With sufficient U233 we can provide a significant long-term solution to fuel our nuclear reactors to produce electricity needed for its development. Linked to the nuclear programme is the availability of fuel, ability to reprocess the spent fuel and manage the wastes. India’s waiver from the Nuclear Suppliers’ Group and its agreement with the global atomic energy body, IAEA, have resulted in limited breakthroughs in the nuclear sector in the last decade and allowed the import of fuel. This paper undertakes a review of the different steps taken by India in the nuclear arena and makes a realistic assessment of its current nuclear power programme. [ABSTRACT FROM AUTHOR]

Three-dimensional structures of a vortical flow field and heat transfer characteristics in a partially blocked 7-pin fuel assembly mock-up of sodium-cooled fast reactor have been investigated through a numerical analysis using a commercial computational fluid dynamics code, ANSYS CFX. The simulation with the SST turbulence model agrees well with the experimental data of outlet and cladding wall temperatures. From the analysis on the limiting streamline at the wall, multi-scale vortexes developed in axial direction were found around the blockage. The vortex core has a high cladding wall temperature, and the attachment line has a low cladding wall temperature. The small-scale vortex structures significantly enhance the convective heat transfer because it increases the turbulent mixing and the turbulence kinetic energy. The large-scale vortex structures supply thermal energy near the heated cladding wall surface. It is expected that control of the vortex structures in the fuel assembly plays a significant role in the convective heat transfer enhancement. Furthermore, the blockage plate and grid spacer increase the pressure drop to about 36% compared to the bare case.

Multigroup cross section (MG XS) generation by the UNIST in-house Monte Carlo (MC) code MCS for fast reactor analysis using nodal diffusion codes is reported. The feasibility of the approach is quantified for two sodium fast reactors (SFRs) specified in the OECD/NEA SFR benchmark: a 1000 MWth metal-fueled SFR (MET-1000) and a 3600 MWth oxide-fueled SFR (MOX-3600). The accuracy of a few-group XSs generated by MCS is verified using another MC code, Serpent 2. The neutronic steady-state whole-core problem is analyzed using MCS/RAST-K with a 24-group XS set. Various core parameters of interest (core keff, power profiles, and reactivity feedback coefficients) are obtained using both MCS/RAST-K and MCS. A code-to-code comparison indicates excellent agreement between the nodal diffusion solution and stochastic solution; the error in the core keff is less than 110 pcm, the root-mean-square error of the power profiles is within 1.0%, and the error of the reactivity feedback coefficients is within three standard deviations. Furthermore, using the super-homogenization-corrected XSs improves the prediction accuracy of the control rod worth and power profiles with all rods in. Therefore, the results demonstrate that employing the MCS MG XSs for the nodal diffusion code is feasible for high-fidelity analyses of fast reactors.

Sodium-cooled reactors (SFR) have always been recognized as one of the most promising candidates for the fourth-generation nuclear systems as announced by the Generation-IV International Forum. In the design of SFR, helical wire-wrapped rod is applied to stabilize the structure of the rod bundle and enhance coolant mixing. Although there has been considerable research on SFR in computational fluid dynamics (CFD), the phenomenon of heat transfer has rarely been paid attention to. This article discovered that there exists reversed heat flux from coolant to wrapped wire, which is contrary to our usual understanding. This phenomenon has not been reported in previous CFD calculations. Hence, a solid heat conduction model is proposed to prove this phenomenon and analyze the heat transfer process. The simulation results show that the wrapping wire embedding depth, the shape of the calculation domain and the physical properties of all components have great influence on the magnitude of the reversed heat flux. The present findings will have strong influence on the temperature field and maximum value of the fuel rod as well as profound reference value for future flow calculation, especially in grid generation and treatment of the junction between the winding wire and fuel rod. [ABSTRACT FROM AUTHOR]

This paper presents floating absorber for safety at transient (FAST) which is a passive safety device for sodium-cooled fast reactors with a positive coolant temperature coefficient. Working principle of the FAST makes it possible to insert negative reactivity passively in case of temperature rise or voiding of coolant. Behaviors of the FAST in conventional oxide fuel-loaded and metallic fuel-loaded SFRs are investigated assuming anticipated transients without scram (ATWS) scenarios. Unprotected loss of flow (ULOF), unprotected loss of heat sink (ULOHS), unprotected transient overpower (UTOP) and unprotected chilled inlet temperature (UCIT) scenarios are simulated at end of life (EOL) conditions of the oxide and the metallic SFR cores, and performance of the FAST to improve the reactor safety is analyzed in terms of reactivity feedback components, reactor power and maximum temperatures of fuel and coolant. It is shown that FAST is able to improve the safety margin of conventional burner-type SFRs during ULOF, ULOHS, UTOP and UCIT.

• APOLLO-3-FR is a deterministic neutronics scheme utilized for comprehensive analysis of power sodium fast reactors/ APOLLO-3-FR shows good agreement with a reference calculation for differential control rod worth • It reveals low discrepancies for main integrated reaction rates, ranging from 0.04% to 0.62%, compared to the reference / APOLLO-3-FR shows acceptable accuracy for further experimental validation studies Design and development of future reactor technologies rely on an intensive use of computational tools for which confidence levels have to be quantified for each targeted quantity of interest. Validation process and uncertainty quantification of such tools must rely on both numerical and experimental data. While zero-power experiments have been widely used in the past, notably for the Sodium-Cooled Fast Reactor (SFR), increased computing power offers the possibility of modeling increasingly complex phenomena. On such a basis, available data from power reactors operation, such as those of the Phenix reactor, can be considered for exhaustive experimental validation of computational tools. In the case of a multi-physic approach, observed phenomena result from multiple factors and origins, each of which may introduce its own modeling bias. Therefore, the decomposition of inherent biases in computational tools for each involved phenomenon is essential to master the accuracy level of any predictive simulation. For neutronics, using deterministic route capabilities offered by the APOLLO3® multi-purpose code, the numerical validation has been performed for each step, from the elementary level (cell, lattice) up to a "Phenix-like" configuration in order to quantify the associated residual biases for the relevant quantities of interest. In the frame of the deterministic approach used here, the APOLLO3®-FR calculation scheme is based on a traditional two-step (lattice/core) assumption. Each step relies on different numerical methods with different parameters whose influence has been identify and quantified. The best set of numerical parameters that lowers the discrepancy compared to the reference calculation tool TRIPOLI-4® (Monte Carlo code) has been determined. It shows that at the 2D lattice scale, before and after homogenization/condensation process, an acceptable accuracy is reached for a further use in 3D core calculation for which validation focuses mainly on the differential control rod worth and reaction rates distribution. Regarding the differential control rod worth, APOLLO3®-FR shows a good agreement with Monte-Carlo reference (<2σ of the TRIPOLI-4® statistical uncertainty). The 239Pu fission and 238U capture rates computed with the deterministic scheme show that the discrepancy with TRIPOLI-4® is within 0.04 % and 0.62 % in global for inner core and outer core zones. These discrepancies are acceptable for further in-depth studies, for instance, fuel depletion or multiphysics analysis, using a more realistic configuration of the Phenix core. [ABSTRACT FROM AUTHOR]

The control rod withdrawal (CRW) transient is one of the most important safety issues for the early design phase of sodium fast reactors (SFR). CRW transients can cause core melting, which can lead to severe accidents. To ensure the integrity of the core, it is crucial to study CRW transients in detail since early design phase of the reactor. Two approaches are used in numerical modeling of transients at CEA. Reference codes (APOLLO3 (Schneider et al., 2016), GERMINAL (Lainet et al., 2019) for example) are used to perform detailed and precise calculations, recently these tools are being coupled to form multiphysics platforms such as described in Pascal et al. (2021); Li et al. (2023). Despite being robust and validated, this type of simulation is not adapted to design phases due to the need in computing resources it requires. The second type of approach, selected in this paper, is to use simplified, parametric tools with a limited validity range. We present here a new numerical tool called MORPHEE, which is capable of simulating CRW transients. MORPHEE is a multiphysics and multiscale tool that includes neutronics, thermal-hydraulics, and fuel behavior. The tool has been developed to provide a comprehensive understanding of the CRW transient and to assess the safety of SFR. It is based on MACARENa (Droin et al. 2020) and OCARINa (Herbreteau et al. 2018) tools. They have been validated for simulation of thermal-hydraulics and fuel pin behavior, respectively. MORPHEE uses finite-differences methods to solve thermal-hydraulics equations, neutron kinetics equations and mechanical equations. A machine learning model (Jeannin et al., 2022) is also used to predict fuel cladding heat transfer through the gap. The simulations have demonstrated the importance of considering the multiphysics and multiscale aspects of the CRW transient to predict the behavior of the reactor during an accident. In conclusion, MORPHEE is a numerical tool that can simulate the CRW transient in SFR. It can be used to study the safety of SFR during the early design phase and to optimize the design to ensure the safety of the reactor. • Multiphysics methodology for Sodium Fast Reactors transients simulation. • Control Rod Withdrawal (CRW) phenomenological-based modeling with MORPHEE tool. • CRW modeling tool to guide design of SFR during the conception phase regarding safety. • Fuel cladding heat exchange coefficient and impact on transients over power. • Neutronics simplified simulation of flux shape deformation. [ABSTRACT FROM AUTHOR]

The SubChannel Module, referred to as SCM henceforward in this document, is a subchannel code within the Multiphysics Object-Oriented Simulation Environment (MOOSE) developed at the Idaho National Laboratory (INL). SCM, is able to model flow through liquid-metal cooled, wire-wrapped fuel pin sub-assemblies, ordered in a triangular lattice. This work extends the existing validation done for these flow types and geometries and introduces validation for the newly implemented capability to model flows within a deformed duct. The subchannel duct deformation modeling approach consisted of reproducing the deformed duct effect on the geometric parameters of the boundary subchannels, i.e., modifying surface area, wetted perimeter and pin to duct gap. Then, the model was validated using experimental data taken from Areva's fuel cooling test facility (FCTF). [Display omitted] • SCM can simulate, sodium-cooled, wire wrapped hexagonal fuel bundles in a deformed duct. • The deformation is modeled by adjusting the geometric parameters of the edge subchannels. • SCM is validated against experimental data obtained from FCTF. • SCM calculations agree well with the experimental measurements. • The effect in the temperature distribution is minimal but observable in the axial pressure-drop. [ABSTRACT FROM AUTHOR]

Properties of molecular clouds (M C s) lying in our Galaxy and their star formation scenarios have been investigated with the help of multivariate unsupervised machine learning techniques concerning several observable parameters. At first, the M C s have been classified into four coherent groups using the standard K-means clustering method. Subsequently, the optimum number of groups has been estimated by applying the Elbow method as well as the computation of Silhouette widths for a robustness check. Later, the properties of the groups are studied through several observable parameters as mentioned along with computed ones e.g. star formation rates (S F R s), virial masses, mass-spectra, dynamical time scales (T d), etc. to get a deeper understanding of the star formation process and dynamical evolution of these clouds. It is found that cluster 1 is suitable for the formation of field stars, binary pairs, or stellar associations, whereas the clouds in cluster 2 and cluster 3 are favorable sites for the formation of Galactic clusters of moderate masses, and cluster 4 may produce massive Galactic clusters as well as a few globular clusters. Surprisingly, for each cluster, clouds at around Galacto-centric radius ∼ 8 kpc, and on the near Galactic plane has a significantly low S F R. These occurrences indicate that the star formation phenomenon has yet not started or the proneness to start in that region. • Properties of molecular clouds (MCs) are studied. • Star formation scenarios in our Galaxy have been investigated. • Multivariate unsupervised machine learning techniques are applied. • The possibility of the formation of field stars, binary pairs, or stellar associations is discussed. • The possibility of the formation of Galactic clusters or Globular clusters is discussed. [ABSTRACT FROM AUTHOR]

This study presents a benchmark analysis of an unprotected loss of flow transient in a sodium-cooled fast reactor at the Fast Flux Test Facility (FFTF), carried out as part of an IAEA coordinated research project. Three codes, namely Serpent (Monte Carlo), DYN3D (3D nodal diffusion) and ATHLET (system thermal hydraulics), were employed in the benchmark exercise. Two distinct modeling approaches were utilized: 1) stand-alone ATHLET with point kinetics (PK) and system thermal hydraulics (TH); and 2) coupled DYN3D/ATHLET with spatial kinetics (SK) and system TH. Neutronics data essential for both approaches were generated using Serpent. The study is organized into three parts. Part I presented a summary of the preparation of neutronics data for PK or coupled SK/TH simulations and included the outcomes of the static neutronics stage of the benchmark. The main focus lay on verifying the cross-section generation method for DYN3D by comparing its results against the reference Monte Carlo solutions obtained with Serpent. Part II provides a detailed description of the ATHLET TH model of the system. This model is thoroughly evaluated by comparing the results with the ANL benchmark solution and experimental data for the transient. Furthermore, a sensitivity analysis is conducted to explore various modeling options and assess their impact on the simulation results. Part III will showcase the transient calculation results using the two modeling approaches. Additionally, an adaptive decay heat model for nodal codes will be introduced. The performance of both modeling approaches will be assessed by comparing their results to the available experimental data. • Detailed description of the ATHLET thermal hydraulics model of the FFTF system. • Direct comparison of our ATHLET model with the ANL's SAS4A model using ANL neutronics data. • Sensitivity analysis to explore various modeling options and assess their impact on the simulation results. [ABSTRACT FROM AUTHOR]

• CEA/IRESNE main recent research outcomes on severe accidents. • Strategy for coherent, generic R&D work for different types of nuclear reactor. • Strategy based on triptych: databases improvement, experimentation and simulation. • Presentation of recent results of this strategy for LWR and SFR. • Research capabilities for a variety of reactor concepts, including Generation IV. In the current context of nuclear revival under the constraint of global warming and an ever-increasing world demand for energy, it is essential to take all precautions in the event of a serious accident during the design phase and to review the severe accident mitigation features of existing plants. The current nuclear revival is accompanied by the development of prototypes based on alternative concepts to light water reactors, such as the Sodium-cooled Fast Reactor (SFR), High Temperature Reactor (HTR), Molten Salt Reactor (MSR), which are also under consideration as reactors of reduced power output, or Small Modular Reactors (SMR). This paper summarizes the main outcomes and achievements of the last five years of experimental and numerical research conducted at CEA/IRESNE into prospective severe accidents within Pressurized Water Reactors (PWR) and SFRs. [ABSTRACT FROM AUTHOR]

Pronghorn-SC is a subchannel code within the Multiphysics Object-Oriented Simulation Environment (MOOSE). Initially designed to simulate flows in water-cooled, square lattice, subchannel assemblies, Pronghorn-SC has been expanded to simulate liquid-metal-cooled flows in triangular lattices, hexagonal subchannel assemblies. For this purpose, the algorithm of Pronghorn-SC was adapted to solve the subchannel equations as they are applicable to a hexagonal wire-wrapped sodium-cooled fast reactor. Cheng–Todreas models for pressure drop and cross-flow models were adopted and a coolant heat conduction term was added. To solve these equations, an improved implicit algorithm was developed robust enough to deal with the numerical issues, associated with low flow and recirculation phenomena. To confirm the prediction capability of Pronghorn-SC, calculations and comparisons with available experimental data of 19- and 37-pin assemblies were performed, as well as other subchannel codes. Finally, a flow blockage modeling feature was added. This capability was validated for both water-cooled square sub-assemblies and sodium-cooled hexagonal sub-assemblies, using experimental data of partially and fully blocked cases.

Objective: To evaluate the safety, efficacy, and feasibility of minipercutaneous nephrolithotomy (MPCNL) under mixture of local anesthetics (MLA) vs spinal anesthesia (SA) for management of large renal stones. Patients and Methods: This study was a prospective randomized controlled study and approved by IRB (REC-FOMBU). A total of 120 consecutive patients who met the inclusion criteria of the study and agreed to sign the informed consent form were randomized to undergo MPCNL under MLA (60 patients) or SA (60 patients). Intra- and postoperative findings including visual pain analogue scale (VAS), operative time, hospital stay, adverse events (AEs), stone-free rate, and related data were recorded. Results: Baseline characteristics and demography included age and gender; stone's site, size, and density were comparable for both groups (p > 0.05). The average VAS scores in the MLA group at 0, 2, 6, 12, and 24 hours were 2.5, 0, 1, 1, and 0, respectively. The corresponding values in the SA group were 2, 1, 2, 2, and 1, respectively, (p < 0.05). The average operation time was ∼1 hour for both groups and the length of hospital stay was 1.5 days for both groups (p > 0.05). Whereas the mean hemoglobin deficit was 1.04% ± 0.54% vs 1.27 ± 0.46 (p = 0.013) and the primary postoperative stone clearance was 93.4% vs 88.3% (p > 0.05), for MLA and SA groups, respectively. Postoperative analgesic consumption and complications were similar in the MLA and SA groups. Conclusion: Single tract MPCNL is feasible under either MLA or SA with comparable stone clearance and AEs. Perioperative VAS was similar and acceptable for both modalities. [ABSTRACT FROM AUTHOR]