Annals of Nuclear Energy最新文献

{"title":"Investigation of coarse mesh acceleration methods for the SN nodal method in unstructured geometries","authors":"Haoxiang Xu ,&nbsp;Youqi Zheng ,&nbsp;Hongchun Wu ,&nbsp;Bowen Xiao","doi":"10.1016/j.anucene.2025.111451","DOIUrl":"10.1016/j.anucene.2025.111451","url":null,"abstract":"<div><div>The S_N nodal method with unstructured nodes is an effective approach for modeling the complicated geometries in solving the neutron transport equation. However, it hits an efficiency bottleneck when triangular nodes are adopted in the modeling. Against this backdrop, this study investigated acceleration methods for the S_N nodal method based on unstructured coarse meshes to address the efficiency problem. To achieve this, the study first proposed a coarse mesh generation algorithm from arbitrary triangular meshes. Then, various CMFD schemes, including pCMFD, odCMFD, and gCMFD, were developed. The proposed method can process regular triangular meshes in structured geometries of hexagonal and rectangle assemblies, as well as arbitrary triangular meshes generated in unstructured geometries using the Delaunay triangulation method. A set of eigenvalue problems with various mesh counts and geometry types was selected to verify the accuracy and evaluate the performance of different acceleration schemes. Results indicated that an acceleration ratio of up to 2–3 can be achieved for different conditions.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111451"},"PeriodicalIF":1.9,"publicationDate":"2025-04-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143800222","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":"Thermal-hydraulics performance and stability two-phase flow using Al2O3 nanofluids in an open natural circulation loop","authors":"Dwi Yuliaji ,&nbsp;Roy Waluyo ,&nbsp;Gatot Eka Pramono ,&nbsp;Muhammad Ganjar Putra ,&nbsp;Nur Rochman Budiyanto ,&nbsp;Shendy Akbar Mariadi ,&nbsp;Sunandi Kharisma ,&nbsp;Ryan Oktaviandi ,&nbsp;Deendarlianto ,&nbsp;Indarto ,&nbsp;Mulya Juarsa","doi":"10.1016/j.anucene.2025.111424","DOIUrl":"10.1016/j.anucene.2025.111424","url":null,"abstract":"<div><div>The thermal-hydraulics performance and stability flow using Al₂O₃ nanofluids in an open natural circulation loop has been investigated. Experiments were conducted with a gradual increase in heating power from 880 <span><math><mi>W</mi></math></span> up to 1350 <span><math><mi>W</mi></math></span>. The working fluid used Al₂O₃ with variations of 0.025 <span><math><mrow><mi>wt</mi><mo>%</mo></mrow></math></span>, 0.05 <span><math><mrow><mi>wt</mi><mo>%</mo></mrow></math></span>, and 0.1 <span><math><mrow><mi>wt</mi><mo>%</mo></mrow></math></span>. The two-phase flow pattern was observed using 1 fps range macro photos. PSD (power spectral density) and DWT (discrete wavelet transform) signal processing were used to analyze the thermal-hydraulics performance and flow stability that occurred during the experiments. The result shows that there are three types of oscillations were found in the observations based on heating power, intermittent oscillation with expulsion-refill-incubation stages, sinusoidal oscillation with expulsion and refill stages, and high subcooling stable flow circulation with only one continuous refill stage. PSD and DWT analysis provided solid agreement between the temperature signal and thermal-hydraulics performance and flow stability.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"218 ","pages":""},"PeriodicalIF":1.9,"publicationDate":"2025-04-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143791319","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":"Source term uncertainty analysis of filtered containment venting scenarios in Nordic BWR","authors":"Sergey Galushin ,&nbsp;Govatsa Acharya ,&nbsp;Dmitry Grishchenko ,&nbsp;Pavel Kudinov","doi":"10.1016/j.anucene.2025.111406","DOIUrl":"10.1016/j.anucene.2025.111406","url":null,"abstract":"<div><div>Nordic Boiling Water Reactors employ filtered containment venting and ex-vessel debris coolability in the deep pool located under the reactor pressure vessel as cornerstones of their severe accident management strategy.</div><div>This paper focuses on the uncertainty analysis of the source term in accident sequences that result in filtered containment venting to the environment using the MELCOR code. The impact of uncertain MELCOR modeling parameters and modeling options on the timing and magnitude of the source term released to the environment has been evaluated in accident sequences initiated by a large break LOCA and SBO.</div><div>The performed simulations illustrate the effect of MELCOR modeling parameters and options on the code’s predictions of severe accident progression, event timing, and the magnitude of the source term released to the environment in different accident scenarios. Furthermore, the results highlight the importance of various retention mechanisms that limit the release of fission products into the environment.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"218 ","pages":"Article 111406"},"PeriodicalIF":1.9,"publicationDate":"2025-04-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143791642","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 novel methodology assessment to study the performance of the physical protection system for enhancing the security of nuclear and other radioactive materials during transport","authors":"Amal Touarsi,&nbsp;Amina Kharchaf,&nbsp;Chakir El Mahjoub","doi":"10.1016/j.anucene.2025.111408","DOIUrl":"10.1016/j.anucene.2025.111408","url":null,"abstract":"<div><div>This paper presents a risk-informed model for securing radioactive materials during transport, integrating a real transport graph to simulate attack probabilities and response effectiveness. The model addresses the potential for adversarial attacks, including theft and sabotage, and evaluates the ability of the Physical Protection System (PPS) to detect, delay, and neutralize threats. Using a real-world transport network, we simulate the most plausible attack paths an adversary might take and assess the success probability of these attacks in relation to key system metrics: detection probability (<span><math><msub><mrow><mi>P</mi></mrow><mrow><mi>d</mi></mrow></msub></math></span>), communication probability (<span><math><msub><mrow><mi>P</mi></mrow><mrow><mi>c</mi></mrow></msub></math></span>), interruption probability (<span><math><msub><mrow><mi>P</mi></mrow><mrow><mi>i</mi></mrow></msub></math></span>), and neutralization probability (<span><math><msub><mrow><mi>P</mi></mrow><mrow><mi>n</mi></mrow></msub></math></span>). The model also considers the arrival time of the response force, as a critical factor influencing the overall security effectiveness. The transport graph represents the network of possible routes, critical locations, and access points, which are used to simulate adversary actions and response dynamics. Each edge in the graph represents a potential attack path, with associated probabilities for adversary success and system response. System effectiveness, <span><math><msub><mrow><mi>P</mi></mrow><mrow><mi>e</mi></mrow></msub></math></span>, is calculated by simulating realistic attack-response scenarios. Results highlight how optimized strategies reduce adversary success while enhancing response efficiency. Case studies demonstrate the model’s utility in securing radioactive materials and improving PPS designs to mitigate theft and sabotage risks.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111408"},"PeriodicalIF":1.9,"publicationDate":"2025-04-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143792447","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":"Impact of molybdenum on iodine chemistry during fission product transport phenomenology","authors":"Muhammad Rizaal ,&nbsp;Kunihisa Nakajima ,&nbsp;Eriko Suzuki ,&nbsp;Shuhei Miwa","doi":"10.1016/j.anucene.2025.111433","DOIUrl":"10.1016/j.anucene.2025.111433","url":null,"abstract":"<div><div>The release of iodine in a case of severe nuclear accident is directly linked to short-term radiological consequences. This concern raises issues in understanding the chemical forms of the transported iodine to devise proper accident management measures/strategies. In contributing to such efforts, this work presents experimental and theoretical approaches to defining the impact of molybdenum as a semi-volatile fission product toward iodine speciation in the gas phase. Given humid atmospheric conditions with different oxygen potentials, the interactions were revealed through the reaction products consisting of both gas and aerosols upon their transport and condensation in the temperature range of 1150–450 K. Through thermodynamic equilibrium calculations where new thermodynamic data of cesium molybdates have been incorporated, the experimental observation was reproduced, hence general interaction mechanism was proposed in this work.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"218 ","pages":"Article 111433"},"PeriodicalIF":1.9,"publicationDate":"2025-04-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143791641","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":"Nuclear data uncertainty propagation in the ARIANE GU3 burnup model using SANDY: Comparison between a FA and a pincell model","authors":"Luca Fiorito ,&nbsp;Lars Engelen ,&nbsp;Federico Grimaldi ,&nbsp;Pablo Romojaro","doi":"10.1016/j.anucene.2025.111423","DOIUrl":"10.1016/j.anucene.2025.111423","url":null,"abstract":"<div><div>Nuclear data uncertainties taken from the general-purpose evaluated libraries JEFF-3.3, ENDF/B-VIII.0 and JENDL-4.0u are propagated through a depletion model of the ARIANE GU3 sample using the SANDY stochastic sampling code combined with the Monte Carlo burnup code SERPENT-2. This approach enabled an accurate characterization of the uncertainty in many nuclide concentrations, for which measurements exist from post-irradiation experiments.</div><div>Stochastic sampling methods for uncertainty propagation in Monte Carlo burnup calculations are notoriously computationally expensive. To address this, the contribution of nuclear data uncertainties to the model response was assessed independently of Monte Carlo uncertainties using a methodology based on conditional estimators. Interestingly, unlike best-estimate values, uncertainty estimates were found to be rather independent of model simplifications. This was demonstrated by comparing uncertainty results for the GU3 fuel assembly model and for a simplified pincell model. The possibility to transpose uncertainties between such models suggests that high assay data accuracy is not strictly necessary for uncertainty analyses. Finally, the variance decomposition analysis revealed gaps in the uncertainty datasets of major nuclear data libraries, leading to an underestimation of total uncertainties in burnup calculations.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"218 ","pages":""},"PeriodicalIF":1.9,"publicationDate":"2025-04-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143791222","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":"Advancements and challenges in small modular lead/lead bismuth eutectic cooled fast reactors: A 30-year overview","authors":"Qiming Yang,&nbsp;Youqi Zheng,&nbsp;Hongchun Wu","doi":"10.1016/j.anucene.2025.111434","DOIUrl":"10.1016/j.anucene.2025.111434","url":null,"abstract":"<div><div>The small modular lead/lead bismuth eutectic cooled fast reactor (SMLFR) has advantages of high inherent safety, low investment risk, long operating cycle and facilitates flexible deployment. Given this, it is regarded as one of the most competitive candidates for the sustainable development of nuclear energy. This review synthesizes advancements and challenges in the design of small modular lead/lead bismuth eutectic cooled fast reactors over the past three decades across different countries. Most of these designs feature compact active cores, extended refueling cycles, and advanced fuel/structural materials, enhancing both economic and safety performance. These innovations provide valuable experiences for the further developments of SMLFRs. However, the practical deployment of SMLFRs faces critical engineering challenges such as material corrosion, thermal shocks and other coolant-related issues. This paper further discusses potential technological advancements to address these challenges and facilitate the commercial application of SMLFR technology.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"218 ","pages":"Article 111434"},"PeriodicalIF":1.9,"publicationDate":"2025-04-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143791640","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":"Subchannel analysis of post-dryout flow and heat transfer in a tight-lattice fuel assembly","authors":"Shuo Chen ,&nbsp;Junsen Fu ,&nbsp;Quanyi Gao ,&nbsp;Yubo Sun ,&nbsp;Yao Xiao ,&nbsp;Tenglong Cong ,&nbsp;Hanyang Gu","doi":"10.1016/j.anucene.2025.111429","DOIUrl":"10.1016/j.anucene.2025.111429","url":null,"abstract":"<div><div>The tight-lattice fuel assembly is a key research area in the design of the high-performance water reactor, of which the thermal–hydraulic characteristics are significantly different from those used in the conventional pressurized water reactor because of its small pitch-to-diameter ratio. Under loss-of-flow accidents, the dryout and post-dryout heat transfer characteristics in the fuel assembly directly impact its safety. However, the relevant studies are limited, restricting the core design and safety analyzation. In this paper, a subchannel code suitable for post-dryout flow and heat transfer in the tight-lattice fuel assembly is developed based on MATRA program by introducing submodels specialized for various flow and heat transfer regimes. By comparing with experimental data of rod bundles, it is confirmed that the developed subchannel code can accurately predict the dry-out location and the heat transfer of mist flow after dryout in the tight lattice fuel assembly. Subsequently, the subchannel code is utilized to analyze the post-dryout heat transfer behavior in the tight lattice fuel assembly. The results indicate that the hydraulic diameter of subchannel significantly influences the flow field distribution. The corner subchannel with a small hydraulic diameter has the lowest flow rate, and thus the vapor void fraction and quality vary fastest along the axial direction. Though the central subchannel with the largest hydraulic diameter has the highest mass flux, but due to the large heating area, the vapor void fraction and quality still increase relatively high. The transverse flow between different subchannels follows the trend from the corner subchannel to the side subchannel, and then to the central subchannel. Due to the high mass flux and high quality in the corresponding subchannel, the dryout occurs earliest on the center rod, resulting in the largest wall temperature. Meanwhile, the spacer grids show stronger effects on post-dryout heat transfer enhancement in each subchannel. The research conclusions could provide reference to the thermal–hydraulic design of the tight-lattice fuel assembly.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"218 ","pages":""},"PeriodicalIF":1.9,"publicationDate":"2025-04-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143791318","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":"Primary containment vessel management during severe accidents of the supercritical water-cooled fast reactor with in-vessel retention","authors":"Kohki Funatsu,&nbsp;Akifumi Yamaji","doi":"10.1016/j.anucene.2025.111438","DOIUrl":"10.1016/j.anucene.2025.111438","url":null,"abstract":"<div><div>The new Primary Containment Vessel management concept for severe accidents of the Supercritical Water-cooled Reactor (SCWR) has been proposed and the characteristics have been clarified through conceptual development of the Japanese fast reactor design (Super FR). The unique once-through direct cycle plant system of Super FR-IVR has enabled combining of In-vessel Retention (IVR) and the compact PCV with the Suppression Chamber (S/C). The plant behaviors during severe accidents were analyzed using MELCOR-2.2. The steam condensation by S/C showed good performance as IVR was effective in limiting the amount of non-condensable gas (H₂ and CO) generation by eliminating Molten Core-Concrete Interaction (MCCI) and ex-vessel metal-water reactions. Furthermore, IVR was effective in controlling the D/W temperature close to saturation temperature of steam. These characteristics are advantageous for the plant system to avoid PCV venting as long as the ultimate heat sink is available or recovered for prolonged duration of the accident progression.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"218 ","pages":"Article 111438"},"PeriodicalIF":1.9,"publicationDate":"2025-04-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143792347","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":"Improved CFD modeling of radial void fraction distribution in flow boiling: A novel approach for wide-ranging operating conditions","authors":"Yujiao Peng,&nbsp;Jianqiang Shan,&nbsp;Junliang Guo,&nbsp;Yudong Zha,&nbsp;Miao Gui","doi":"10.1016/j.anucene.2025.111445","DOIUrl":"10.1016/j.anucene.2025.111445","url":null,"abstract":"<div><div>Accurate prediction of void fraction distribution in flow boiling is of great significance for the thermal–hydraulic analysis and safety of nuclear power plants. Computational Fluid Dynamics (CFD) methods have consistently proven to be effective for boiling simulation. Due to the complexity of boiling phenomena, the prediction of radial void fraction distribution in tubes tends to be challenging, thus hindering a clear understanding of the underlying mechanism. In this work, with the aim of achieving the void fraction prediction over a wide range, the DEBORA experiment operated with R12 was simulated within the pressure range of 1.46–3 MPa and the mass flux range of 1000–5000 kg·m⁻²·s⁻¹. The closures of the two-phase equation were compared and the applicability of R12 was discussed. Meanwhile, sensitivity analyses were conducted on the models for bubble departure diameter, lift force, and turbulent dispersion force. The Kocamustafaogullari model was selected for calculating bubble departure diameter. According to the cross-sectional average void fraction, the appropriate selection criteria for the effective turbulent Prandtl number model was recommended. An abnormal upward trend of void fraction near the wall was observed for <em>G</em> ≥ 3000 kg·m⁻²·s⁻¹ due to the small or even negative relative velocity rather than the lift force coefficient model. A new lift force coefficient model relevant to pressure, mass flux, and void fraction has been proposed. The new CFD model combination demonstrated a good agreement with the experiment, providing an effective tool for analyzing the radial void fraction distribution.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111445"},"PeriodicalIF":1.9,"publicationDate":"2025-04-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143792446","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}