Annals of Nuclear Energy最新文献

{"title":"Mathematical modelling of bubble formation at top-submerged nozzles placed inside a quiescent liquid","authors":"Soumya Sarkar ,&nbsp;Nirvik Sen ,&nbsp;K.K. Singh","doi":"10.1016/j.anucene.2025.111281","DOIUrl":"10.1016/j.anucene.2025.111281","url":null,"abstract":"<div><div>Bubble size is an important variable that determines the efficiency of air induced liquid mixing, which, due to its passive nature, is commonly used in processes involving radioactive species. In majority of cases, in air induced mixing devices used in radiochemical plants, air is introduced through top-submerged nozzles. This necessitates fundamental understanding of the phenomenon of bubble formation at top-submerged nozzles placed inside a liquid pool. In this study, a mathematical model is developed by carrying out force balance to estimate the diameter of the bubbles generated at the tip of a top-submerged nozzle submerged in a quiescent liquid. The model is validated with the experimental data. The parametric analysis carried out with the developed model provides detailed understanding of the effects of geometric, operating parameters and properties of the liquid medium on bubble formation phenomenon.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111281"},"PeriodicalIF":1.9,"publicationDate":"2025-03-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143579629","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Global sensitivity analysis using Gaussian process surrogate models & impact of thermophysical properties uncertainties on the steady-state operation of molten salt reactors","authors":"Fadel M. Nasr ,&nbsp;Mauricio Tano ,&nbsp;Piyush Sabharwall ,&nbsp;Yousry Azmy","doi":"10.1016/j.anucene.2025.111304","DOIUrl":"10.1016/j.anucene.2025.111304","url":null,"abstract":"<div><div>Measurement uncertainties in the thermophysical properties of molten salts affect the predicted steady-state operation of molten salt reactors (MSRs). We aim to quantify that impact using three MOOSE-based multiphysics models of MSRs: the molten-salt fast reactor (MSFR), the molten-salt reactor experiment (MSRE), and the molten-chloride reactor experiment (MCRE). For each model we perturb the fuel salt’s thermophysical properties using Gaussian and uniform distributions, then we analyze the impact on the statistical distributions of various quantities of interest (QoIs) that are representative of the steady-state operation of MSRs. These QoIs can be grouped into two categories: those related to energy and those related to flow. Our results demonstrate that while the resulting uncertainties depend on the considered QoI, they are predominantly influenced by density uncertainty. Specifically, for the MSFR, heat capacity and density have the greatest impact on uncertainty. In the MSRE, the key contributors are heat capacity and dynamic viscosity, while in the MCRE, thermal conductivity, dynamic viscosity and density are the most significant sources of uncertainty. In general, for energy-related QoIs in the MSFR and MSRE cases, the observed QoI distributions were Gaussian regardless of the distribution chosen for the input thermophysical properties, i.e., uniform or Gaussian. For the flow-related QoIs, the distribution of the QoI is dependent on the prior distribution. However, interestingly, multimodal distributions are observed for the energy-related QoIs of MCRE, suggesting a possible bifurcation behavior in the multiphysics model. Refining the mesh for the MCRE model causes the multimodal distributions to disappear suggesting that the bifurcation is spurious, nonphysical. This article establishes a more general framework for uncertainty quantification in MSRs that accounts for the nonlinearity of the underlying models.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111304"},"PeriodicalIF":1.9,"publicationDate":"2025-03-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143563278","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"A higher order implicit stair-tailored scheme for the modified Burgers’ equation","authors":"V.P. Shyaman,&nbsp;A. Sreelakshmi,&nbsp;Ashish Awasthi","doi":"10.1016/j.anucene.2025.111284","DOIUrl":"10.1016/j.anucene.2025.111284","url":null,"abstract":"<div><div>In the quest of constructing a user-friendly handy algorithm for the numerical solutions of the modified Burgers’ equation, an adaptive finite point method is carefully orchestrated with no compromise on precision. The proposed algorithm has been designed so as the localized behavior of the analytic solutions is innately inherited to the numerical solutions. The original equation first undergoes a linearization post which the algorithm revolves around an implicit 4-point stair-shaped framework. Within this structure, the solution of the linearized equation at each node in the advanced temporal level is expressed as the linear combination of the nodal solutions at the current and previous temporal levels. Apart from being conditionally stable, consistent, converging, and rapid the method above and beyond replicates the exact solutions on coarse meshes even when the kinematic viscosity close in to zero.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111284"},"PeriodicalIF":1.9,"publicationDate":"2025-03-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143563279","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Simulation of the effect of partial and total blockage in a subchannel on flow features distribution and interactions in sodium-cooled 5 × 5 rod bundles","authors":"Ayodeji A. Ala ,&nbsp;Zhu Feng ,&nbsp;Liu Junyan ,&nbsp;John Njoroge ,&nbsp;Bin Ye","doi":"10.1016/j.anucene.2025.111327","DOIUrl":"10.1016/j.anucene.2025.111327","url":null,"abstract":"<div><div>Subchannel flow obstruction accident is one of the safety considerations for sodium-cooled fast reactors. A computational fluid dynamics simulation was performed to study the flow of liquid sodium in a 5 × 5 fuel assembly with partial and entire blockage of a sub-channel. A situation with no blockage was used as a baseline for further investigation into following cases with blockages. The effect of a 75 % and total obstruction in a subchannel on the flow characteristics and heat transfer in the fuel assembly was studied from −5 hydraulic diameters before the blockage to 50 hydraulic diameters after the blockage and compared. The temperature surge of 10 K due to flow through the spacer in the model without blockage increased to 14 K and 24 K in the models with partial obstruction and total blockage, respectively. The blockage-induced resistance coefficient, turbulence kinetic energy distribution, and temperature fluctuation in the adjacent subchannels can be monitored to detect the blockage.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111327"},"PeriodicalIF":1.9,"publicationDate":"2025-03-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143563276","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Coupled reactor physics and thermal-hydraulics analysis of an SCW-SMR reactor concept – Part II: Improvement of the concept and sensitivity analyses","authors":"Tamás Varju,&nbsp;Csenge Antók,&nbsp;Péter Mészáros,&nbsp;Boglárka Babcsány","doi":"10.1016/j.anucene.2025.111221","DOIUrl":"10.1016/j.anucene.2025.111221","url":null,"abstract":"<div><div>In the frame of the EU ECC-SMART project, the pre-conceptual design of a water-cooled small modular reactor operating at supercritical pressure (SCW-SMR) with seven heat-up stages is under development and assessment. The first part of the two-part paper presented the concept design, the models developed at BME in the Apros thermal-hydraulics system code and in the Serpent 2 Monte Carlo reactor physics code, the developed coupling methodology of these models, and the results of the first coupled calculations. This second part of the paper outlines the approaches adopted by the BME research team to improve the original design, primarily to reduce the high cladding temperatures. The effects of the increased mass flow rate and the introduction of inhomogeneous inlet orifices are thoroughly investigated and presented. Finally, a comprehensive sensitivity analysis has also been performed to demonstrate the robustness of the proposed design.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111221"},"PeriodicalIF":1.9,"publicationDate":"2025-03-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143696884","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Cross-Verification and validation of steady-state OPAL research reactor neutronic models using OpenMC and MCNP5 Codes: A foundation for advanced coupled simulations","authors":"Djahid Lababsa ,&nbsp;Hakim Mazrou ,&nbsp;Djalal Hamed ,&nbsp;Abdelkader Aouaichia ,&nbsp;Hamza Hasnaoui","doi":"10.1016/j.anucene.2025.111324","DOIUrl":"10.1016/j.anucene.2025.111324","url":null,"abstract":"<div><div>This study develops, verifies, and validates high-fidelity steady-state neutronic models of the OPAL research reactor using OpenMC and MCNP5. The models were developed based on a consistent methodology, utilizing identical input data, geometric modeling, and cross-section libraries, minimizing discrepancies from external factors, ensuring a solid foundation for subsequent code-to-code comparisons.</div><div>Verification and validation were conducted through comparisons with results from SERPENT and MCNP6, as well as experimental data. Our results showed strong agreement with the experimental data, including control rod worth calculations within 3% and axial peak factor differences averaging 2% for MCNP5 and 3% for OpenMC. Differences in isothermal temperature and void coefficients were approximately 1 pcm/°C and 7 pcm/%, respectively. Deviations from SERPENT and MCNP6 for few key parameters were within acceptable limits, consistent with reference studies, and attributed to variations in modeling, code differences, and Monte Carlo statistical uncertainties.</div><div>Finally, a comprehensive code-to-code comparison of key neutronic parameters, such as reactivity, flux distributions, kinetic parameters, and power distribution, showed excellent agreement between OpenMC and MCNP5, with relative differences under 1% for most parameters.</div><div>Overall, this study confirms the reliability of the developed models for simulating the OPAL reactor and establishes OpenMC as a viable alternative to MCNP for reactor physics simulations, laying the foundation for future coupling with thermal–hydraulic models.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111324"},"PeriodicalIF":1.9,"publicationDate":"2025-03-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143563277","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Improving neutron economy of a Lead-Bismuth Eutectic-Cooled reactor using Modified-CANDLE axial fuel shuffling","authors":"Nina Widiawati ,&nbsp;R.Andika Putra Dwijayanto ,&nbsp;Fitria Miftasani ,&nbsp;Nuri Trianti ,&nbsp;Anni Nuril Hidayati ,&nbsp;Maryam Afifah ,&nbsp;Ratna Dewi Syarifah ,&nbsp;Zaki Su'ud","doi":"10.1016/j.anucene.2025.111328","DOIUrl":"10.1016/j.anucene.2025.111328","url":null,"abstract":"<div><div>LWRs do not effectively utilize natural uranium as fuel, harnessing less than 1% of its energy potential. Furthermore, initial studies show that conventional fast reactors require higher uranium enrichment and fuel reprocessing, which are expensive technology. The Modified-CANDLE strategy enables the direct use of natural uranium without enrichment or reprocessing. This study implements modified-CANDLE axial fuel shuffling in a lead–bismuth eutectic (LBE)-cooled fast reactor to examine its impact on reactor performance compared to radial fuel shuffling. The calculations use the SRAC code and the JENDL 4.0 nuclear data library. This research evaluates the effective multiplication factor (k-eff), power density, power peaking factor (PPF), and neutron leakage over a 15-year fuel cycle. The research indicates that axial fuel shuffling is more efficient at sustaining the reactor’s reactivity than radial shuffling. Therefore, axial fuel shuffling can improve neutron economy compared to radial fuel shuffling.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111328"},"PeriodicalIF":1.9,"publicationDate":"2025-03-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143550075","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"A review of the scientific contributions by Barry Ganapol","authors":"P. Ravetto ,&nbsp;P. Saracco","doi":"10.1016/j.anucene.2025.111253","DOIUrl":"10.1016/j.anucene.2025.111253","url":null,"abstract":"<div><div>Barry Ganapol has been an outstanding personality within the transport theory and the nuclear reactor physics communities throughout the past 50 years. He is universally recognized as one of the scientists who has given relevant contributions to the fields of mathematics and computations in nuclear science and he has shown an uncommon capability to export this knowledge to different areas, with a rare attitude towards interdisciplinarity. He has obtained significant achievements in the development of methods for the solution of the linear transport equation, in reactor physics and in radiative transport analysis. He is widely known and appreciated for having established a wide variety of accurate benchmarks for many problems within these fields. This paper reviews the scientific production of Barry Ganapol, highlighting what are believed to be his most important scientific accomplishments.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111253"},"PeriodicalIF":1.9,"publicationDate":"2025-03-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143550074","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Integrated thermal power measurement in the modified STACY for the performance inspections","authors":"Shouhei Araki ,&nbsp;Eiju Aizawa ,&nbsp;Takahiko Murakami ,&nbsp;Yu Arakaki ,&nbsp;Yuta Tada ,&nbsp;Yutaka Kamikawa ,&nbsp;Kenta Hasegawa ,&nbsp;Tomoki Yoshikawa ,&nbsp;Masato Sumiya ,&nbsp;Masakazu Seki ,&nbsp;Junichi Ishii ,&nbsp;Kazuhiko Izawa ,&nbsp;Daiki Iwahashi ,&nbsp;Shigeki Shiba ,&nbsp;Satoshi Gunji","doi":"10.1016/j.anucene.2025.111323","DOIUrl":"10.1016/j.anucene.2025.111323","url":null,"abstract":"<div><div>Japan Atomic Energy Agency (JAEA) has modified the Static Experimental Critical Facility (STACY) to a heterogeneous system using fuel rods in order to obtain criticality characteristics of fuel debris. Thermal power measurement was required for the calibration of a power meter system in the modified STACY in order to conduct a series of performance inspections and operation. We measured the thermal power using the modified activation method that combined MVP and PHITS code because neutron flux distribution cannot be measured experimentally. Four operations were conducted for the thermal power measurement. The power meter was calibrated by using three operational data and tested with one operational data. It was found that the calibration of the power meter system was successfully agreed within 3% with the results measured by the activation method.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111323"},"PeriodicalIF":1.9,"publicationDate":"2025-03-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143550076","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Sensitivity analysis of parameters and optimization of produced entropy in a nuclear reactor with equilateral fuel rod arrangement and variable angle turbulators","authors":"Abdullah A.A.A. Al-Rashed,&nbsp;Abdulwahab A. Alnaqi,&nbsp;Jalal Alsarraf","doi":"10.1016/j.anucene.2025.111310","DOIUrl":"10.1016/j.anucene.2025.111310","url":null,"abstract":"<div><div>This paper investigates the produced entropy (ENT), including thermal ENT, fluid loss, total ENT, and the Bejan number (Be) in a nuclear reactor with equilateral fuel rod arrangements. The fuel rods are placed within tubes that have diameters varying from 3.1 to 10.1 cm. The rods are vertically positioned in the reactor at a fixed distance, and nanofluids are used to cool them. Turbulator (TUR) blades with angles varying from 0 to 6 degrees and heights from 0 to 13 cm are employed in the nanofluid flow. To enhance the analysis, machine learning methods are used for sensitivity analysis of parameters and optimization of outputs. The results of this study indicate that increasing the tube diameter and the height of the TUR blades leads to an increase in fluid loss ENT, while changes in blade angle have a minimal effect on fluid loss ENT. Increasing the tube diameter and blade length significantly increases thermal ENT and total ENT. Blade angle also has an effect, with thermal ENT decreasing as the angle increases. Changes in blade height, angle, and tube diameter containing the fluid flow result in a 1.8 % variation in the Be. The results demonstrate that tube diameter, blade length, and TUR blade angle are all important parameters in determining the Be in the reactor. An increase in tube diameter significantly raises the Be, while an increase in blade length decreases it. The angle of the blades also affects the Be, with an increase in angle leading to a decrease in the Be.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111310"},"PeriodicalIF":1.9,"publicationDate":"2025-03-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143550073","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}