bluuebunny/crossref_metadata_2025_split huggingface dataset:
Experimental Investigation of Heat Transfer to Supercritical Pressure R134a in Artificially Roughened Tubes
Authors: Wiltschko Fabian , Sipova Monika , Vit Jan , Cheng Xu 
Published: November 2024

Abstract
The environment in supercritical water-cooled reactors (SCWR) is highly corrosive. With increasing time of operation, an oxide layer grows on the surfaces of the fuel rod cladding, which increases the roughness of their surface. The aim of this study is the investigation of the effect of the roughness due to corrosion on the heat transfer to a fluid at supercritical pressure conditions. The drastic variations in the thermophysical properties of SCW can lead to a deterioration or an enhancement of the heat transfer. The correct prediction of the heat transfer deterioration, which can cause sharp temperature peaks on the fuel rod cladding is a big challenge. Until the current day, most experimental investigations are conducted, using hydraulically smooth heat transfer surfaces. In the present investigation, the effect of the surface roughness is investigated systematically with experiments, using R134a as a working fluid. In the first step, heat transfer experiments are conducted in a hydraulically smooth directly heated tube, covering a wide range of operating conditions. In the second step, the experiments were repeated, using a tube with a rough inner surface. Based on the results from the two test tubes, the effect of the surface roughness on heat transfer is evaluated.
Improved Athermal Recombination Corrected DPA Model for Displacement Damage Cross Section Calculation
Authors: Yin Wen , Zu Tiejun , Cao Liangzhi 
Published: November 2024

Abstract
Accurate evaluation of material irradiation damage is the basis for safe and economical operation of nuclear facilities. Displacement damage is an important component of material irradiation damage, and the level of displacement damage in materials is usually described by the displacement damage rate. The estimation of displacement damage rate requires the prior knowledge of displacement damage cross section and displacement function model is the key to the displacement damage cross section calculation. Traditional displacement damage function models have significant errors in the calculation of the displacement damage function near the average atomic displacement threshold energy because they use step function approximately for the calculation of the number of displaced atoms. In order to improve the accuracy of displacement damage cross section calculation, an improved ARC-DPA (iARC-DPA) model is proposed, which uses the power function model near the average atomic displacement threshold energy and original ARC-DPA model in the other energy part. The iARC-DPA model has been used in the calculation of displacement damage cross sections for different incident particles, including electron, neutron and charged particles. The displacement damage cross sections of different particles are calculated and compared with the experimental data. Numerical results show that the iARC-DPA model largely improves the accuracy of displacement damage cross section calculation.
Considerations on Current Methodologies for the Assessment of Engineering Correlations for Heat Transfer at Supercritical Pressures
Authors: Kassem Sara , Pucciarelli Andrea , Ambrosini Walter 
Published: November 2024

Abstract
The design of industrial equipment operating at supercritical pressures, especially in innovative nuclear reactors like Supercritical Water Reactors (SCWRs) at 25 MPa, demands accurate modeling of convective heat transfer. The prediction of fuel surface temperature is vital for safety and regulatory approval of new reactor concepts.
Over the years, numerous engineering correlations have been developed based on experimental data for various fluids at supercritical pressures. Some of these correlations claim high accuracy, but their performance often falls short when applied to different datasets, particularly under deteriorated heat transfer (DHT) conditions. DHT is a complex phenomenon sensitive to boundary conditions, challenging traditional formulations like the Dittus-Boelter equation with property ratio corrections. Even more sophisticated formulations have not significantly improved accuracy.
Within the EU ECC-SMART Project, a comprehensive review of these correlations was performed, evaluating their performance using two diverse datasets covering a wide range of boundary conditions. Great care was applied in avoiding common pitfalls in assessing correlation accuracy, such as the inappropriate use of experimental wall temperature data, which can bias the results and mislead about the most critical variable to be predicted by a correlation when used in a system code.
The paper presents sample results from this assessment, highlighting the potential and limitations of existing formulations. It also addresses cases where multiple solutions for wall temperature arise, even with simple formulations. We consider correlations both in their original form and with optimized coefficients, shedding light on possible enhancements. This study offers insights into the current state-of-the-art in heat transfer modeling for supercritical fluids and suggests directions for future research.
Inquiry on the Effect of Heating Structures in the Prediction of Flow Stability at Supercritical Pressures
Authors: Betti Alessio , Pucciarelli Andrea , Ambrosini Walter 
Published: November 2024

Abstract
Flow stability is an important concern in the design of apparatuses containing fluids with heating and cooling sections. Fluid density changes, even when they are minimal as in singlephase thermosyphon loops, are often responsible for the occurrence of flow oscillations, which are mostly unwanted while striving for stable operating regimes. Nuclear reactors may be prone to these phenomena, something that poses problems in view of the stability of operation requested, inter alia, by the General Design Criterion 12 in Appendix A to the USA 10 CFR 50 law. A design conducive to stable operation and an early detection and suppression of oscillations is therefore in order.
In the frame of a third IAEA Coordinated Research Project on Supercritical Water Reactors (SCWRs), aiming to close some of the open issues in this technology, the problem of the effect of heating structures on the prediction of flow stability at supercritical pressure is now targeted for being attentively scrutinised. In front of the scarce experimental data available that sometimes provide counterfeiting information with respect to the one proposed by models, a varied modelling approach must be used, searching for possibly overlooked ingredients in the currently adopted numerical recipes. In the present paper, different modelling techniques have been adopted, ranging from the use of 1D time-domain and linear in-house programmes, to system codes and CFD. With reference to some experimental data and predictions appeared in literature, an attentive discussion is carried out on the possible reasons for discrepancies between observations and predictions, aiming to contribute to clarify this open issue.
Investigation in Design Technology of Load Follow Without Changing Soluble Boron of Hualong One
Authors: Huaiming Cui , Zhonghao Ning , TongXian Liu , ChenLin Wang , Tian Ya Li , Yun Cai 
Published: November 2024

Abstract
At present, Hualong One adapted Mode-G operation and control strategy, which needs to adjust soluble boron concentration periodically in daily load follow and unavoidably make some waste water generation. In order to enhance the advancement and competitiveness of Hualong One, we have proposed the design of Mode-C operation and control strategy to replace Mode-G operation and control strategy in the Hualong One improvement project, which can choose single variable automatic control mode or dual variable automatic control mode according to the demand of load change. One of the most notable advantages of Mode-C is that the core can realize typical daily load follow without changing soluble boron in dual variable automatic control mode. This paper investigates the design technology of load follow without changing soluble boron of Hualong One in Mode-C operation and control strategy. The main design contents include control rods set design, control strategy investigation and theoretical calculation. The theoretical practice shows that the design technology proposed by this paper can effectively help Hualong One realize typical daily load follow without changing core soluble boron, which can do notable contribution to reductions in waste water generation.
Inverse Uncertainty Quantification for Subchannel Code With PSBT Experimental Benchmark
Authors: Luo Hanyu , Liu Xiaojing 
Published: November 2024

Abstract
Inverse uncertainty quantification analysis for subchannel programs has been conducted to decide uncertainty distribution of important model parameters based on the modified Markov Chain Monte Carlo (MCMC) algorithm using the PSBT void fraction experiment. The posterior probability distribution expression for input parameter uncertainty can be derived using Bayesian principles, typically representing a proportional relationship. However, obtaining the direct normalization coefficient proves challenging, prompting the use of the MCMC algorithm to address this issue. Given the substantial external forward program calculations associated with stochastic methods, the Back-propagation Neural Network (BPNN) is employed to construct a surrogate model, serving as a substitute for intricate forward calculations. To significantly enhance the precision of the surrogate model, an adaptive approach based on relative entropy minimization is implemented to densify training sample points. The uncertainty of key input model parameters (slip ratio and turbulent mixing coefficient) was quantified, revealing that the uncertainty of the slip ratio is the primary source of void fraction uncertainty. After obtaining information on input parameter uncertainty, forward uncertainty analysis based on input uncertainty was conducted. The resulting uncertainty band for void fraction was obtained under a tolerance limit of 95%-95%. The results indicate that the uncertainty band effectively envelops the experimental data for void fraction. Using the statistical mean of parameter uncertainty obtained, the baseline values were calibrated. The predicted results of the model correction values were found to be more accurate than those of the baseline predictions.
Conceptual Design and Feasibility Analysis of a Modular Supercritical CO2 Fast Reactor Core
Authors: Lai Guihua , Wang Tian , She Ding , Wang Hong 
Published: November 2024

Abstract
Small Modular Reactor (SMR) is a type of prospective nuclear reactor which has gained extensive attention due to its flexibility, economics, and safety. This paper presents an innovative design of a fast reactor core cooled by supercritical CO2 (SCO2), which can operate at 50 MWth for at least 20 years without refueling. A tube-in-duct structure using beryllium oxide (BeO) as matrix material is proposed. Specifically, the heat generated by the fuel rods is conducted by BeO matrix to adjacent coolant channels, and the SCO2 removes the heat out of the core. Reflectors, shieling, and control rods are applied for operation modification and safety considerations. To enhance the inherent safety, the reactor used double-layer SiC particles with uranium nitride (UN) as fuel. These particles are randomly dispersed in lead contained in the middle of the fuel rods, while the lead offers a better conduction performance. Partial 238U is substituted by minor actinides (MA) to harden the neutron spectra. In the feasibility analysis, both neutronic and thermal-hydraulic calculations are performed under normal condition. Results show that when the reactor is operated at 50 MWth for 30 years, the burnup is about 98.123 MWD/kgHM and the maximum power peaking factor is 1.57 in the radial direction and 1.2 in the axial direction. The inlet/outlet temperature of SCO2 is 660K/873K, while the maximum temperature inside the core holds below 1600 K, which is under the temperature limits. This work may provide a reference for the design of small modular reactor in the future.
Study on Radial Thermal-Hydraulic Characteristics of Once-Through Steam Generator of Sodium-Cooled Fast Reactor
Authors: Liu Yapeng , Wang Bo , Zhang Dalin , Feng Zhenyu , Li Xinyu , Tian Wenxi , Qiu Suizheng , Su Guanghui 
Published: November 2024

Abstract
In sodium-cooled fast reactor, once-through steam generator (OTSG) is an important heat transfer component, and it is also one of components that is prone to failure. The safe and stable operation of this equipment is crucial to the safety of sodium-cooled fast reactors. The steam generator has a sizable number of heat exchange tubes. Variations of mass flow rates in distinct radial zones on the primary side of the steam generator are caused by structural variations in the steam generator, which in turn cause variations in flow and heat transfer along the radial direction. Further thermal stresses are produced as a result of these variations in the fluid and structure’s temperatures. As a result, more detailed modeling of the primary side of the steam generator is required. An essential technique for a detailed simulation of the primary side of the once-through steam generator is computational fluid dynamics (CFD). However, CFD method is time-consuming and difficult to apply to long-term transient analysis. Therefore, in order to calculate the long-term transient operating conditions, and considering the radial thermal-hydraulic difference on the primary side of the steam generator, a radial parallel multi-channel model is developed to analyze the radial thermal-hydraulic difference of the once-through steam generator. This model can use multiple parallel channels to model the primary and secondary sides of the heat exchange tubes in different areas of the steam generator. The lateral momentum exchange between the heat exchange tubes is not considered in the model, but the energy exchange between different radial regions can be considered in the form of heat conduction. First, the model was verified using the transient test conditions carried out on the once-through steam generator test facility (SGTF) in India. The calculation results were consistent with the test results, and the deviation was within 10 K. The test results and calculation results in different radial areas were compared, and the program’s ability to calculate radial thermal-hydraulic differences was verified. Finally, the test conditions of the steam generator prototype of the China Fast Reactor 600 (CFR600) were analyzed. By considering the difference in radial flow distribution, the results of single channel model and radial multi-channel model are compared, and the impact of the radial thermal-hydraulic difference of the once-through steam generator on the simulation results was analyzed. The calculation results show that the radial multichannel model can effectively analyze the flow/heat transfer differences of the steam generator in different radial areas. It is verified that the model can analyze the radial thermal-hydraulic difference of the steam generator, and can provide reference data for the rapid thermal-mechanical coupling analysis of the steam generator.
Containment Depressurization Rate Requirements for Design Basis Loss-of-Coolant Accidents
Authors: Zheng Hua 
Published: November 2024

Abstract
Currently passive containment cooling system (PCCS) of HPR1000 is used in design extension conditions. PCCS uses loop-type heat pipes with evaporators inside the containment and condensers and a large water storage pool outside the containment. One of HPR1000 potential optimizations is using PCCS to mitigate design basis accidents, but there is an issue of containment depressurization rate. Section 6.2.1.1.A of Standard Review Plan (SRP) requires the containment pressure should be reduced to less than 50% of the peak calculated pressure for the design basis loss-of-coolant accident within 24 hours after the postulated accident. Due to the working principles of PCCS and available layout space inside the containment, it ‘s difficult for PCCS to meet the containment depressurization rate requirement. Safety requirements of nuclear safety regulations, guides, standards and utility requirements documents are analyzed to trace back to the provenance and its technical basis. Design features for meeting containment depressurization rate requirements of typical light water cooled reactors (LWR) are analyzed and compared. A combination of increasing the containment design pressure and increasing heat transfer area of PCCS is proposed for HPR1000. Loss of coolant accident is analyzed for HPR1000, which demonstrates that after increasing the heat transfer area and the design pressure of the containment, HPR1000 can meet the containment depressurization rate requirement.
Droplet Evaporation Characteristics on High-Temperature Porous Surfaces for Cooling Fuel Debris
Authors: Yuki Kohei , Horiguchi Naoki , Yoshida Hiroyuki , Yuki Kazuhisa 
Published: November 2024

Abstract
Fuel debris at the Fukushima nuclear power station is typically cooled under immersion. However, significant drop in water level results in coolant contact with high-temperature, porous fuel debris. In such scenarios, rapid cooling is essential, yet the thermal behavior at the liquid-solid interface, including capillary phenomena, is not well understood. This paper presents basic research evaluating the evaporation characteristics of droplets upon contact with metallic porous media featuring small pores under 1 mm. We conducted experiments using bronze or stainless steel porous media with pore diameters of 1, 40, or 100 μm to derive lifetime curves for droplets. Our findings indicate that Leidenfrost effect is mitigated on porous surfaces as the vapor can escape through the pores. Moreover, in bronze porous media, as the temperature increases, oxide film with a fine structure facilitates capillary action. In contrast, the low wettability of stainless steel porous media prevents capillary action, inhibiting droplet absorption and dispersion into the pores. Consequently, rapid cooling via the capillary action is unlikely if the fuel debris shares similar characteristics with steel porous media.
Research on CFD Calculation and Analysis Technology of Porous Media in Sodium Cooled Fast Reactor Core
Authors: Sun Yuchen , Chen Guangliang , Yin Xinli , Qian Hao , Tian Yizhi , Zhang Senyong , Li Menglai 
Published: November 2024

Abstract
The composition of sodium cooled fast reactor fuel components is relatively simple, without too many complex structures. Therefore, it is necessary to carry out thermal hydraulic simulation of the entire core and large watershed. However, Due to the complexity of the existing CFD refined methods in modeling and grid partitioning, the long calculation time lead to the inconvenience of the real time analysis. Therefore, the porous medium method is used for CFD analysis in the entire core and inter-wrapper flow of sodium cooled fast reactors. At the same time, the conjugate heat transfer method is used to simulate the fuel assembly box wall and separate the flow inside and between the boxes for research. The research results indicate that the porous media method can greatly simplify the difficulty of modeling and the number of grid divisions. Through comparative analysis and verification, it is found that the influence of inter-wrapper flow on the flow inside the core is limited. Therefore, in actual modeling, the inter-wrapper flow, as well as components with extremely low power flow rates such as neutron sources and reflector layers, are ignored. Only fuel components are analyzed, while other components are ignored. On this basis, flow partitioning was applied to the core, and the results were compared with those without partitioning. The results showed that flow partitioning promoted the flattening of the core outlet temperature. Finally, the calculation results will be compared with experimental data to demonstrate the effectiveness of the method using porous media. The calculation has achieved sub real-time and has good engineering application value.
Transition Mechanism of Supercritical Carbon Dioxide Between Gas-Like and Liquid-Like States
Authors: Liu Minyun , Han Yansong , Huang Yanping 
Published: November 2024

Abstract
In recent years, the supercritical carbon dioxide power cycle has made great breakthroughs and has been applied to new energy fields such as solar heat and waste heat recovery. In the field of nuclear energy, the supercritical power cycle has the advantages of high efficiency, compact structure, and good compatibility with active media, so it has broad application prospects in the new generation of nuclear energy systems combined with high-temperature gas cooled reactors, sodium-cooled fast reactors, and even fusion reactors. Supercritical fluids are typically regarded as homogeneous, continuous, indistinguishable, single-phase substances without undergoing phase transitions. However, recent research has demonstrated the existence of a transition mechanism between gas-like and liquid-like structures within supercritical fluids. Considering that this structural transition may lead to the mixing of nitrogen and supercritical carbon dioxide within the pressurizer, this paper devised a straightforward visual experiment. The experiment investigated the characteristics of the interface between supercritical carbon dioxide and nitrogen under different pressure and temperature conditions. The results revealed that, under the same pressure, as the temperature rises, the interface between supercritical carbon dioxide and nitrogen evolves from being clear to gradually becoming blurry and ultimately vanishing. Under a pressure of approximately 7.7 MPa, nitrogen and supercritical carbon dioxide will completely mix at temperatures above 31 °C. To address the significant interference caused by this issue in thermal-hydraulic experiments, this paper also introduces an effective pressurizer design.
Numerical Analysis of Three-Dimensional Flow Field Impact Under Strong Transient Conditions in Advanced Reactors
Authors: Zhao Jian , Zhang Jing , Wu Yingwei , Su Guanghui , Tian Wenxi , Qiu Suizheng 
Published: November 2024

Abstract
As a representative of Generation IV nuclear energy systems, sodium-cooled fast reactors (SFR) may cause the liquid sodium flowing through the core to boil after a Core Disruptive Accident (CDA), which then forms a high-energy vapor cavity expansion to push the liquid sodium above, causing an impact failure to the main vessel structure in a very short period of time, which is a safety hazard. In this study, the computational analysis of the process of high-energy vapor cavity expansion pushing liquid sodium impacting on the top of the reactor vessel was carried out from the full scale to derive the effects of high-energy impact on the inner wall and upper components of the reactor vessel after the accident. The results show that after the CDA, the sodium-cooled fast reactor shielding lid is subjected to a great impact pressure at the center of the cork, while the covering argon gas at the top is compressed, resulting in a rapid increase in the internal pressure of the reactor vessel within a short period of time, and extrusion of the vessel inner wall. Through this study, the flow field distribution characteristics inside the reactor vessel in the late stage of the core disruptive accident are grasped, and the distribution of the impact pressure on the top shielding cover at the full reactor scale in the core disruptive accident is obtained, which can provide technical and result references for the studies related to the CDA of pool-type sodium-cooled fast reactors (PSFR).